• 방사선산업학회지
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• 제17권4호
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• pp.333-344
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• 2023

To validate the numerical model used in the study of deep disposal of spent nuclear fuel, we selected benchmark cases and performed numerical model validation. We selected the DECOVALEX-THMC Task D_THM1 FEBEX Type benchmark case, which was conducted from 2003 to 2007. We analyzed the thermal-hydraulic (TH) behavior using the finite element program CODE_BRIGHT and verified the results against previous studies. The temperature results were similar to the results of DECOVALEX-THMC Task D. The saturation results showed a similar trend to the results of DECOVALEX-THMC Task D, but the time to reach full saturation was different.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1762-1776
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• 2024

A spent nuclear fuel (SNF) analysis module for the Vodo-Vodyanoi Energetichesky Reactor (VVER) was developed and validated in this study. This advancement expands the application area of the existing nodal diffusion code, RAST-V, and reduces the need for additional code during 3D core simulations for SNF analysis, leading to increased efficiency in simulation time. RAST-V uses Lagrange interpolation and a power correction factor derived from the Bateman equation to bypass the re-depletion calculations, which are used to solve the microdepletion chain. This approach improved the efficiency of analysis. To mirror the conditions during the 3D core simulations, the module used history indices related to the moderator temperature, fuel temperature, and boron concentration. The module can predict 1620 isotopes. This paper presents the validation of this isotope inventory prediction and the application of burnup credit. The VVER analysis module was validated using 28 samples discharged from the Novovoronezh-4. Most isotopes were within 10 % of the boundaries of the measurements. This study successfully offers verification results using VVER benchmarks and discusses the application of burnup credit using a VVER-440 cask.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.603-616
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• 2009

The past two decades were mainly devoted to model validation and computer code verification against global corium experiments, code application to reactor situations, and investigation of the role of melt properties in steam explosion energetics. Corium data were essentially provided by JRC-Ispra in the FARO and KROTOS facilities and by KAERI in the TROI facility. Verification of code applicability to reactor situations was performed essentially in the frame of the international OECD/SERENA programme. The paper makes a synthesis of the findings made during the above-mentioned period and expresses a personal view of the author with respect to the progress made and expected for the resolution of the steam explosion issue for light water reactors.

• 제어로봇시스템학회논문지
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• 제15권11호
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• pp.1144-1149
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• 2009

This paper deals with the time synchronization problem over SpaceWire links. SpaceWire is a standard for high-speed links and networks between spacecraft components, which was invented for better, cheaper, faster on-board data handling in spacecraft. The standard defines Time-Code for time distribution over SpaceWire network. When a Time-Code is transmitted, transmission delay and jitter is unavoidable. In this paper, a mechanism to remove Time-Code transmission delay and jitter over SpaceWire links is proposed and implemented with FPGA for validation. The proposed mechanism achieves high resolution clock synchronization over SpaceWire links, complies with the standard and can be easily adopted over SpaceWire network.

• 반도체디스플레이기술학회지
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• 제11권1호
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• pp.67-71
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• 2012

The optical properties of PDP, such as the transmittance and reflectance, were analyzed with 3D optical code. Three different ITO-less structures in the front panel are examined. In the assembled panel study, the test 1 structure shows 16.6% and 10.2% higher reflectance than the structures of tests 2 and 3, respectively. In order to check the validation of the simulation result, three 7.5-inch test panels having the same geometry and property are fabricated as simulation models. The calculated reflective properties are compared to the measured data from real panels. The relative difference extracted from the simulation and measurement methods is less than 4.9% and are well matched.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.713-718
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• 1998

Henry-Fauske critical flow model was incorporated into TRAC-PF1 to correct some errors in the original TRAC-PFI critical flow model. Henry-Fouske mode1 was numerically implemented and tested against steady-state steam-water experimental data. The model was incorporated into TRAC-PFI and code assessment against Marviken Critical Flow Tests 15 and 24 was carried out. Calculations using RELAP5/MOD3 were also made for comparison. Ten cases were calculated each test and sensitivity study on nodalization as well as critical flow or model was performed Stand-alone numerical model test and code assessment were done for verification and validation of code modification. Calculation results show that the modified version of TRAC-PF1 has a capability to model critical flow correctly in various conditions.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2002년도 추계 학술대회논문집
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• pp.28-34
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• 2002

A computational study of supersonic flow around lateral jet controlled missile has been performed. For this study, three dimensional Navier-Stokes code(AADL3D) has been developed. Spalart-Allmaras one equation turbulence model has been implemented on the AADL3D code for relatively rapid computational time. For the validation of developed code, AADL3D, pressure distributions on an ogive-cylinder body has been compared with experimental data. Also, the shock structure of sonic jet on the flat plate in the supersonic flow field has been compared with experimental flow visualization result to see the analysis capability of freestream-jet interaction case. A case study has been performed through comparing the normal force coefficient and the moment coefficient of missile body for several jet flow conditions. Current results will be used to the optimum design of a lateral jet controlled missile.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2017년도 제48회 춘계학술대회논문집
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• pp.1000-1002
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• 2017

본 연구에서는 내재적 이중시간 전진기법을 이용한 비정상 Navier-Stokes 코드를 개발하였다. 내재적 이중시간 전진기법은 가상시간에 대한 새로운 잔류항을 도입하는 개념으로 비정상 잔류항에 실시간 미분항을 더한 잔류항을 가상시간으로 푸는 기법이다. 비정상 코드 검증 방법으로 Stokes 2nd 문제인 '주기성 하모닉 진동을 하는 평판 위의 층류 유동'을 해석하였다. 계산된 속도분포와 마찰계수를 방정식 이론적 해와 비교한 결과 매우 근접한 수치해를 얻을 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.21-27
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• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.